This invention relates to the art of controlling a water chemistry of nuclear reactor water, and more particularly to a sensor suitable for measuring the water chemistry of the reactor water as well as to a method of controlling the water chemistry of the reactor by the use of such a sensor.
It is thought that intergranular stress corrosion cracking (hereinafter referred to as "stress corrosion") of a structural material of a reactor occurs when three factors, that is, the composition and the stress of the material and the water chemistry, are all in undesirable conditions. Conventionally, in order that the reactor can be operated sufficiently safely from the viewpoint of stress corrosion, the carbon content of the structural material (particularly, stainless steel (SUS304)) of the reactor has been lowered, or a heat treatment has been applied to the structural material to relieve a residual stress. Thus, the conventional measures have coped with two (i.e., the material and the stress) of the three factors in the stress corrosion. Recently, in a boiling water-type reactor (BWR), hydrogen injection has been attempted in order to reduce dissolved oxygen in reactor water to improve the water chemistry (i.e., the third factor), as disclosed in Japanese Patent Unexamined Publication No. 57-3086.
In this prior art, a feed-water system, disposed downstream of a condenser of a primary cooling system of the BWR, is provided with a hydrogen injection device which is disposed upstream of a feed-water pump, and hydrogen injected from this hydrogen injector device is recombined with oxygen produced as a result of radiolysis of the water at a reactor core, in order to reduce the dissolved oxygen concentration in the various parts of the primary cooling system including a recirculation system.
A major part of the injected hydrogen shifts to a vapor phase under a boiling two-phase flow condition of the reactor core, and is discharged or emitted to the outside of the primary cooling system via a main steam piping and a turbine. When hydrogen is injected, hydrogen becomes greatly excessive relative to the oxygen being discharged simultaneously, and therefore it is necessary that upstream of the offgas oxygen-hydrogen recombiner, oxygen or the air of such an amount as corresponding to the excess of hydrogen should be supplied. In many cases, the oxygen concentration in the reactor water is measured usually by providing a sampling system in a reactor purification system.
It has been confirmed that stainless steel (SUS304), intentionally increased in sensitivity to corrosion, will not be subjected to stress corrosion if the dissolved oxygen is reduced to about 20 ppb.
Outside a pressure vessel, the effects of the hydrogen injection can be confirmed by measuring the dissolved oxygen concentration, using a water chemistry measuring system; however, there is no effective means for confirming the environment within the pressure vessel. Confirming means now used on a trial basis include sampling the water by the use of a neutron instrument pipe, and inserting a crack growth monitor into an instrument pipe so as to monitor the environment within the reactor core indirectly through the crack growth rate of a test piece made, for example, of stainless steel (SUS304). These measures have the following disadvantages:
(1) Since the sampling water, while flowing through the neutron instrument pipe, is reacted and recombined in accordance with the intensity of radiation exposure in the reactor, the measured value obtained outside the reactor does not represent the true environment within the reactor.
(2) In the crack growth evaluation device, electric current is applied from an external power source to both ends of a crack growth test piece as prestressed, and an electric potential between two specific points of the test piece is measured to determine the length of the crack. In order to ensure the precision of the measurement, it is necessary to use two pairs of potential-measuring points. Therefore, a total of six wires per test piece must be led in from the exterior of the reactor, and the number of test pieces is substantially limited to one for each neutron instrument pipe.
Therefore, in order to improve the evaluation precision, a plurality of sensors are needed. Further, other materials (e.g. inconel) than stainless steel (SUS304) are used as the structural material of the pressure vessel, for which evaluation is also required. And besides, a complicated software depending on a computer is needed for the evaluation of the crack growth behavior.
The hydrogen injection has another limitation. Namely, radioactive nitrogen (N-16) usually dissolved in the water in the form of nitric acid is reduced by the injected hydrogen into gas, so that the dose rate of the turbine system and hence that of the site boundary increase. In an example of the actual installation, it has been reported that the dose rate increased to about five times at the maximum with increase of the amount of the injected hydrogen. There is a tendency that the dose rate is kept constant below a certain threshold value of the hydrogen concentration, and abruptly increases above such a threshold value. Therefore, the amount of the hydrogen has an upper limit, and with respect to the hydrogen injection operation in the BWR, it is necessary to relieve the environment by injecting the hydrogen of an amount less than the upper limit.
Japanese Patent Application Nos. 62-259711 and 63-154767 describe that a radiation dose rate in the primary steam system can be reduced by injecting nitrous acid or nitric oxide gas. With such method, basically, the dissolved oxygen concentration can be reduced without increasing the radioactive nitrogen concentration in the turbine system. If the injected amount is excessive, the dissolved oxygen concentration increases although the radioactive nitrogen concentration decreases, and also the electric conductivity of the water becomes high, and a profile is formed in the primary system, which results in difficulty of the control.